BVY 13-022, Revision of Technical Specification Bases Pages

Revision of Technical Specification Bases Pages

ML13088A047
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	03/26/2013
From:	Wanczyk R Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BVY 13-022
Download: ML13088A047 (6)

Text

Entergy Nuclear Operations, Inc.

Vermont Yankee a Entergy 320 Governor Hunt Rd Vernon, VT 05354 Tel 802 257 7711 Robert J. Wanczyk Licensing Manager BVY 13-022 March 26, 2013 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Revision of Technical Specification Bases Pages Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28

REFERENCES:

1. Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station -

Issuance of Amendment RE: Change to Suppression Chamber-Drywell Leak Rate Test Surveillance Frequency (TAC NO. ME7928),"

NVY 13-009, dated January 30, 2013

2. Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station -

Issuance of Amendment RE: Rod Worth Minimizer Bypass Allowance (TAC NO. ME7927)," NVY 13-008, dated January 30, 2013

Dear Sir or Madam:

This letter provides revised Vermont Yankee Nuclear Power Station (VY) Technical Specification (TS) Bases pages. The TS Bases were revised in conjunction with Amendments to Renewed Facility Operating License DPR-28 issued in References 1 and 2.

These changes, processed in accordance with our TS Bases Control Program (TS 6.7.E), were determined not to require prior NRC approval. The revised Bases pages are provided in for your information and for updating and inclusion with your copy of VY TS. No NRC action is required in conjunction with this submittal.

There are no new regulatory commitments being made in this submittal.

Should you have any questions concerning this submittal, please contact me at 802-451-3166.

Sincerely,

,RJW/plc]

BVY 13-022 / Page 2 of 2

Attachment:

1. Revised Technical Specification Bases Pages cc: Mr. William M. Dean Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Mr. Richard V. Guzman, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 08C2A Washington, DC 20555 USNRC Resident Inspector Entergy Nuclear Vermont Yankee, LLC 320 Governor Hunt Rd Vernon, Vermont 05354 Mr. Christopher Recchia, Commissioner VT Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601

BVY 13-022 Docket No. 50-271 Attachment 1 Vermont Yankee Nuclear Power Station Revised Technical Specification Bases Pages

VYNPS BASES: 3.3 & 4.3 (Cont'd)

2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system. The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

3. In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer (RWM) restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. With the RWM inoperable during a reactor startup or shutdown, the operator is still capable of enforcing the prescribed control rod sequence.

However, the defense in depth is reduced since a single operator error can result in violating the control rod sequence. Therefore, control rod movement must be immediately suspended except by scram.

Alternatively, startup may continue if at least 12 control rods have already been withdrawn or a reactor startup with an inoperable RWM was not performed in the last calendar year. Once these conditions have been verified by either control room indication or control room logs, the RWM function can be performed manually following a second check of compliance with the prescribed rod sequence by a second licensed operator or other qualified member of the technical staff.

The RWM may be bypassed under these conditions to allow continued operation or shutdown. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 17% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.

4. Refer to the "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-24011-P-A, (the latest NRC-approved version will be listed in the COLR).

5. The Source Range Monitor (SRM) system provides a scram function in noncoincident configuration. It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.

Amendment No. 2-5, -74, 4-44, BzV 99 !1!, By9_4_ 0, 422-4, 2-3-3, 255 90

VYNPS BASES: 3.3 & 4.3 (Cont'd)

6. The action statement for TS 3.3.B.6 requires that the plant be placed in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the required actions of TS 3.3.B.1 through 3.3.B.5 are not satisfied. This ensures that all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The allowed completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based upon operating experience to reach HOT SHUTDOWN from full power in an orderly manner and without challenging plant systems.

Amendment No. L&-, 7--3, 4-4&, VY 99 !1!, BY- 4, 2-2-4, 244, 255 90a

VYNPS BASES: 4.7 (Cont'd)

Every 18 months, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by at least 1 psi with respect to the suppression chamber pressure and held constant. The 2 psig set point will not be exceeded. The subsequent suppression chamber pressure transient (if any) will be monitored with a sensitive pressure gauge. If the drywell pressure cannot be increased by 1 psi over the suppression chamber pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed. If the drywell pressure can be increased by 1 psi over the suppression chamber the rate of change of the suppression chamber pressure must not exceed a rate equivalent to the rate of leakage from the drywell through a 1-inch orifice. In the event the rate of change exceeds this value then the plant will be shut down, if operating, the source of leakage will be identified and addressed before power operation is resumed. Two consecutive test failures, however, would indicate unexpected primary containment degradation; in this event, increasing the frequency to once every 9 months is required until the situation is remedied as evidenced by passing two consecutive tests.

The drywell-suppression chamber vacuum breakers are exercised in accordance with Specification 4.6.E, following termination of discharge of steam into the suppression chamber from the safety/relief valves and following any operation that causes the vacuum breakers to open. This monitoring of valve operability is intended to assure that valve operability and position indication system performance does not degrade between refueling inspections.

When a vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights are designed to function as follows:

Full Closed 2 White - On (Closed to <0.050" open)

Open 2 White - Off

(>0.050" open to full open)

Experience has shown that a weekly measurement of the oxygen concentration in the primary containment assures adequate surveillance of the primary containment atmosphere.

B. and C. Standby Gas Treatment System and Secondary Containment System Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 0.15 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leakage tightness of the reactor building, and performance of the standby gas treatment system. The testing of reactor building automatic ventilation system isolation valves in accordance with Technical Specification 4.6.E demonstrates the operability of these valves. In addition, functional testing of initiating sensors and associated trip channels demonstrates the capability for automatic actuation. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system performance capability.

Amendment No. 1-2-, 4-4-7, 2-49, 254 169